1. Field of the Invention
The present invention relates to a high Fe-containing zirconium alloy composition having excellent corrosion resistance and a preparation method thereof.
2. Description of the Prior Art
Nuclear fuel claddings, spacer grids, and nuclear reactor core structures, which are used in nuclear fuel assemblies in nuclear power plants, become brittle due to high-temperature/high-pressure corrosive environment and neutron irradiation, and suffer reduction in mechanical properties due to a phenomenon of corrosion product growth, and thus alloy compositions thereof are very important. Accordingly, zirconium alloys having low neutron absorption cross sections and excellent mechanical strength and corrosion resistance have been widely applied in pressurized water reactors (PWRs) and boiling water reactors (BWRs) for several decades. Among zirconium alloys developed to date, Zircaloy-2 (comprising 1.20-1.70 wt % tin, 0.07-0.20 wt % iron, 0.05-1.15 wt % chromium, 0.03-0.08 wt % nickel, and 900-1500 ppm oxygen, the balance being zirconium) and Zircaloy-4 (comprising 1.20-1.70 wt % tin, 0.18-0.24 wt % iron, 0.07-1.13 wt % chromium, 900-1500 ppm oxygen, and up to 0.007 wt % nickel, the balance being zirconium), which comprise tin (Sn), iron (Fe), chromium (Cr) and nickel (Ni), are most widely used.
However, in order to increase the economic efficiency of nuclear reactors, high burnup/extended cycle operations have recently been adopted, in which the cycle of nuclear fuel is extended to reduce the life-cycle cost of nuclear fuel. As the cycle of nuclear fuel is extended, the period during which nuclear fuel reacts with high-temperature and high-pressure water and steam is extended. For this reason, when Zircaloy-2 and Zircaloy-4 are used as materials for nuclear fuel claddings, a problem occurs in that the phenomenon of corrosion due to nuclear fuel becomes severe.
Accordingly, there is an urgent need to develop materials, which have excellent corrosion resistance to high-temperature and high-pressure water and steam, and thus can be used in nuclear fuel assemblies for high burnup/extended cycle operations. Thus, many studies focused on the development of zirconium alloys having improved corrosion resistance have been conducted. Herein, because the corrosion resistance of zirconium alloys is greatly influenced by the kind and amount of additional elements, processing conditions, heat treatment conditions, and the like, it is particularly important to establish optimal conditions that show excellent corrosion resistance.
With respect to major patents relating to nuclear fuel assemblies for high-burnup/extended cycle operations, which were registered after the middle of the 1980s, zirconium alloys mostly comprise iron, which can improve corrosion resistance, even when it is added in trace amounts. Also, in Fe-containing zirconium alloy compositions, it is a general tendency to increase the amount of added iron and to add other elements that have an effect of improving corrosion resistance. That is, zirconium alloys for high burnup/extended cycle nuclear fuels essentially contain a high concentration of iron, and optimal preparation processes thereof are established such that the zirconium alloys exhibit excellent performance.
U.S. Pat. No. 5,648,995 discloses a method for preparing a zirconium alloy comprising 0.005-0.025 wt % iron, 0.8-1.3 wt % niobium, 0.16 wt % and less of oxygen, 0.02 wt % and less of carbon, 0.012 wt % and less of silicon, and the balance of zirconium. This patent attempts to improve creep resistance by restricting the content of iron to within a range of very low values.
U.S. Pat. No. 5,112,573 discloses a process for preparing a zirconium alloy, having iron content higher than that of U.S. Pat. No. 5,648,995, and comprising 0.07-0.14 wt % iron, 0.5-2.0 wt % niobium, 0.7-1.5 wt % tin, 0.03-0.14 wt % nickel or chromium, 0.022 wt % and less of carbon, and the balance of zirconium.
U.S. Pat. No. 5,125,985 and U.S. Pat. No. 5,266,131 relate to a manufacturing process, in which a “late stage” beta-quenching process is performed during the cold processing of a zirconium alloy having the same composition as that of U.S. Pat. No. 5,112,573. These patents attempt to improve creep resistance and corrosion resistance.
U.S. Pat. No. 5,940,464 discloses an alloy composition having iron content about 20 times higher than that of U.S. Pat. No. 5,648,995, and comprising 0.02-0.4 wt % iron, 0.8-1.8 wt % niobium, 0.2-0.6 wt % tin, 30-180 ppm carbon, 10-120 ppm silicon, 600-1800 ppm oxygen, and the balance of zirconium, as well as a preparation process thereof. This patent attempts to improve corrosion resistance and creep resistance.
U.S. Pat. No. 5,211,774 discloses an alloy composition comprising 0.2-0.5 wt % iron, 0.8-1.2 wt % tin, 0.1-0.4 wt % chromium, 0-0.6 wt % niobium, 50-200 ppm silicon, 900-1800 ppm oxygen, and the balance of zirconium, as well as a preparation process thereof. This patent attempts to reduce the variations in corrosion resistance according to the absorption of hydrogen by changing the content of silicon in the alloy and the difference of the process.
U.S. Pat. No. 5,254,308 discloses an alloy composition that maintains its mechanical properties due to a decrease in the content of tin, and comprises 0.4-0.53 wt % iron, 0.45-0.75 wt % tin, 0.2-0.3 wt % chromium, 0.3-0.5 wt % niobium, 0.012-0.03 wt % nickel, 50-200 ppm silicon, 1000-2000 ppm oxygen, and the balance of zirconium. The above patent, the iron/chromium ratio was controlled to be 1.5, the added amount of niobium was determined by the added amount of iron which affects the hydrogen absorption property of the alloy. Further, the added amount of nickel, silicon, carbon, and oxygen was determined to provide excellent corrosion resistance and strength.
U.S. Pat. No. 5,560,790 discloses an alloy composition comprising 0.3-0.6 wt % iron, 0.5-1.5 wt % niobium, 0.9-1.5 wt % tin, 0.005-0.2 wt % chromium, 0.005-0.04 wt % carbon, 0.05-0.15 wt % oxygen and 0.005-0.015 wt % silicon. In this patent, the interparticle distance between intermetallides (Zr(Nb,Fe)2, Zr(Fe,Cr,Nb) and (Zr,Nb)3Fe) is 0.20-0.40 μm, and the intermetallides are at least 60 volume percent of the total amount of ferriferous intermetallides.
In Europe Patent No. 198,570, the content of niobium in a binary alloy consisting of zirconium-niobium is limited to 1.0-2.5 wt %. This patent also discloses that the temperature of heat treatment performed during a process for preparing the alloy can lead to improved corrosion resistance.
U.S. Pat. No. 5,125,985 discloses an alloy comprising: 0.5-2.0 wt % niobium; 0.7-1.5 wt % tin; and 0.07-0.28 wt % of at least one element selected from among iron, chromium and nickel. Also, this patent discloses that the creep resistance of the material can be controlled by subjecting the material to various treatment processes.
As described above, there have been continued efforts to improve the corrosion resistance and mechanical properties of zirconium alloys used as materials for nuclear fuel assemblies in nuclear powder plants. However, zirconium alloys having further improved corrosion resistance, which can secure the integrity of nuclear fuel in high burnup/extended cycle operations, are continually required in consideration of the tendency toward high burnup/expended cycle operations, in which the cycle of nuclear fuel is expanded to increase the economic efficiency of power plants, and the target burnup is increased.
Accordingly, the present inventors have conducted many studies to improve accelerated corrosion phenomena, which are most problematic when nuclear fuel claddings, spacer grids and structures, made of zirconium alloys, are used in high burnup/extended cycle operations. As a result, the present inventors have found that a zirconium alloy composition, containing 0.5-1.0 wt % iron and prepared using varying kinds of additional elements through an optimized preparation process, has excellent corrosion resistance compared to the prior Zircaloy alloys, thereby completing the present invention.